ML20210A313
| ML20210A313 | |
| Person / Time | |
|---|---|
| Site: | General Atomics |
| Issue date: | 04/30/1987 |
| From: | Asmussen K GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation |
| References | |
| 67-1061, NUDOCS 8705050061 | |
| Download: ML20210A313 (23) | |
Text
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... 4 l k GA Technologies Inc.
SAN D EGO AUFORNIA 92138 (619) 455-3000 i
April 30,1987 !
67-1061 Office of Iniclear anactor Regulation U.S. Itaclear Regulatory Oceanission .
leishington, D.C. 20555 !
Attenticn Document Control Desk Rabject: Docket No. 50-163: License No. IH 7; Request for i Technical specification diange (3 copies)
Dear Sir GA Technologies Inc. (m) has need to amend its subject reactor
'mchnical Socifications to allow the uninterrupted continuation of a government (Dot) research programi on thermionic (i direct conversion) devices now in progress. The DDR program at GA to test '
the radiation effects on thermionic devices has been in operation -
since January, 1985. 'this program has been very successful and will soon rest.h the point where the technical specification limit on -
! irradiation time enast be extenried if the testing is to continue as l l desired. Currently, Section 10.2.6, item (d), of the subject !
l technical gecification limits the irradiation time for any one direct I conversion device to not more than 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. les need to extend the irradiation time beyond 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> in order to further test :
these devices. 'therefore, we hereby request that section 10.2.6(d) of E cur subject technical specification be amended to read: ;
"10.2.6 (d) 'the irradiation time for any one device shall not exceed 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.'
'the first of the devices under test is expected to reach 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> in september,1987. It is hoped that our request- to extend the si duration of the irradiation time to 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> will be granted by that time.
S.
S 15 'the justification for amending the subject technical specification to extend the irradiation time to 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> is given in the attached safety analysis. 'the safety analysis considers both the fission product inventory and the radiation effects on those components that ig assure contairunent, both primary. and secondary, of the thermionic devices. 'the results of these analyses indicate that no sipiificant increase in the gasecus fission products of concern will result.
$ Marther, althoudi the radiation effects after 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> will cause
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some degredation in the contalment materials, for the purposes involved in the present tests, this degredation will not affect its ability to contain fission products and will not diminish the safety of the system. Please see the attachment for these details.
CA concludes that there is no increase in the likelihood or conse-quences of hazards involved with the requested technical specification mendment.
We look forward to receiving the requested technical specification amendnent at your earliest convenience. Should you have any questions regarding this awlication, please contact me or Dr. W. L. Whittemore at (619) 455-2823 and (ti19) 455-3277, respectively. We thank you for your early attention given to our request.
Ehclosed is a check for $150 to cover the administrative fee.
Very truly yours, Mud d'. k Keith E. Aaraussen, Manager Licensing, Safety and Nuclear Capliance KEA/mk Attachments: 1. Safety analysis dated April 24, 1987
- 2. Check for $150 SIATE & CALIET)RNIA )
)ss GXMY T SAN DIEGO )
On this the 30th day of April,1987, before me, ha'wr des ,
the undersigned Notary Public, personally appeared K. E. Amussen, Manager, Licensing, Safety and Nuclear C a pliance, proved to me on the basis of satisfactory evidence to be the person whose name is subscribed to the within instrument, and acknowledged that he executed it. WrINESS my hand and official seal.
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Attachment to 67-1061 April 24, 1987 SMTIT AleLYSIS &
DOE 1EIRUGEIC TEST MCGRAM Backaround Discussion 4
One of the major goals of the present DOE test progra for thermionic devices is to evaluate the dimensional stability of the enitter that is heated by nuclear fuel. An understanding of this stability is essential to the design of higher efficient and long-lived thermionic cells because it effects the size of the gap between emitter and collector. For increased efficiency, this gap should be anall. But if the emitter swells as a result of radiation h===,
! the original gap must be made larger to acc % te this swelling. W e current I thermionic tests are being conducted to monitor emitter distortion as a func-tien of fuel burnup and confirm our understanding of the underlying mechanians.
We present faily of thermionic devices was designed and constructed not to produce electric power but to present controlled circumstances in which to study enitter distortion. Consequently, the gaps between emitter and collector in the present devices are larger than needed to acWte the growth expected in 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of irradiation. In fact, this gap is well able to
, accMte the growth expected in more than 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of irradiation.
We growth of the enitter dianeter is carefully monitored through neutron radiographic (NIO measurenents made about every 1500-2000 hours of irradiation.
h e conditions for NR imaging are such that precise dimensions can be obtained fra the radiographs. Analyses of these results show that the observed growth and that predicted on the basis of applicable theory agree very well. De 4
ocmiputer code used to predict the growth contains parameters which can be adjusted if appropriate to yield improved interpretations of existing data and better forecasts of future data. After following the gradual growth for nearly 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, we are confident that no danger exists to cepletely close the l enitter-to-collector gap of any of the cells as a result of the ongoing growth expected up to 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Even if a gap closure were to result fra unexpectedly large growth, no danger to the public could result because no Page 1
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fission products would be released. S e only effect would be that the device
.being tested would have significantly reduced temperature and would no longer
[ be able to operate as a thermionic device.
- In the following, we evaluate (1) the effect on fission product inventory j fra increasing the irradiation time fra 20,000 to 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, and, (2) the l
! effect of radiation on the critical elements of the containment system. he j latter are mainly Type 304 stainless steel, the welds made from Type 309 i
l welding rod, and the hard vacuum seal fabricated fra ceramic and metal.
! Pission Pr*t Inventory After 40.000 Nrmes IrradimHrwi i
}
} In this section, we will demonstrate that only one gaseous fission product-nuclide (Kr-85, T1/2 = 10.2 yr) will have increased activity as a result of
]l increasing the duration of the tests from 20,000 to 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. It will also .
l 1
be demonstrated that this increase will contribute a negligible increase
( (0.0964) to the concentration of fission products of concern at the site j .
doundary.
i j 2e fission product inventory in a direct conversion (thermionic) device j is directly related to the total thermal power in the device due to fissioning l of U-235. %e desired thermal power is set by thermionic requirements. $e thermionic device may contain more than one individual thermionic converter.
l j ne gaseous fission products fra each converter may be contained in individual i
fission gas traps or vented to a causon line and a single gas trap. To j evaluate the consequences of an accidental release of fission prochets, we have
=====d the leest conservative of these two choices and ====d that the
] l l cabined volume was released. In determining the accidental release of fission i products to the reactor room the following assumptions have been made:
1
! 1. S e only halogens to escape the device would be I and Br.
l i I
- 2. 100% of the be mine inventory would escape.
i
- 3. lot of the iodine inventory would be allowed to escape the device.
j Furthermore, in calculating the release from the reactor rom, it was I
%2
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===w that at least 50% of the iodine inventory plated out inside r l.
the reactor rom. Additional data and analyses were presented in an l; l
earlier license application
- to indicate that the percentage of the i
I-131 inventory available for release into the secondary contairunent
~4 '
is 2.4 x 10 4 his information demonstrates that the use of a 10% i j release is excessively conservative and could be justifiably reduced :
l signiticantly.
! t f 4. 100% of the xenon and krypton would be allowed to escape. If special !
! sorbing material is used within the thermionic cell contaiment (such i as activated charcoal), then much less than 33% of these nuclides 1
would escape.
l i 'm evaluate the effects of increasing the irradiation time fra 20,000 f
j hours to 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, we have prepared Table 1. ;
I
! nis ===arizes the data for the highest operating power level (3.7510t) +
)
contemplated for any of the irradiation tests to be conducted in the present j program. Listed there for the-appropriate gaseous fission products are 7
- (1) the inventory of curies within the device at 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />; (2) the curies j releasable fra the reactor rom in the event of an accidental release;. (3) the
! corresponding half lives; and (4) the fraction of MFC at the site boundary.
l Fra the Table we see that the maximan accidental release of the gaseous 1
! fission products fra one device (containing up to 100 grams of U-235) results
! in only 18.8 percent of MFC at the site boundary. Moreover, it must be l eughasized that extremely conservative estimates of the release fractions were l a == =ad in these calatlations, especially for the iodines which are the
! majority contributors to MEC. ;
i
)i f *The details of calculations taade for the Mark III (F100) reactor at GA
)
Technologies are essentially the suas as for the present Mark F reactor and can l be found in the safety Evaluation for the Direct Conversion Devices, dated December 21,1!r10, Gh-9622 autaitted as port of the 11 cones application for the I GA Mark III reactor (now doomenissioned).
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i_,,,_-..____-,,_.
'mB2 1. Gaseous Fission Products Released to Reactor Rom and to Site Bomdary at 350 m. after 20,000 HR Irradiation (3.75 KN Opera-tion) with Stated Assumptions.*
Curies
- Curies T Relaanable Fraction of l/2 From MPC** MPC at Nuclide Inventory Reactor Ro m pc/ml Site Boundary, ff
-10 I-131 91.4 8.04d 4.6 1x10 0.1617
~9 132 151.6 2.29hr 7.6 3x10 0.00025
-10 133 204.7 20.8hr 10.2 4x10 0.01953
~9 134 237.5 52.6m 11.9 6x10 0.00008
~9 135 184.4 6.58hr 9.2 lx10 0.00242
-8 Br-82 3.9 35.3hr 3.9 4x10 0.0012
~7 Kr-85 47.7 10.72y 47.7 1x10 0.00009 85.2 85.2 210
-8 0.00023 87 76. min 88 115.4 2.84hr 116.4 2x10
-8 0.00066
~
Xe-131m 0.8 11.92d 0.8 4x10 0.000008 133m 4.7 2.19d 4.7 3x10
-7 0.000027
~7 133 204.7 5.25d 204.7 3x10 0.0020
~7
- 135 171.9 9.09hr 171.9 lx10 0.000624 Ifg g = F = 0.188
- I-54 release; Br-100% release; Kr 1004 release; Xe-100% release.
1
- 10CFR20 (1/24/84),168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> week, unrestricted area.
1 Page 4
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l The results presented in Table 1 can be used to estimate the effect on the
} public safety of increasing the irradiation time frca 20,000 to 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. [
f Except for Kr-85 (T1/2 = 10.72 yr), all other nuclides have such shorter half [
] ' lives and reach saturation well before 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. None of these has l
- j. increased activity after 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.- Twenty thousand hours (~2.3 yrs) is :
l short enough compared to the 10.72 half life of Kr-35m that the increase of its j activity is still nearly linear with irradiation time. Therefore, in 40,000 j hours the accumulated activity will nearly double fran 47.7 Curies to about 1 95.4 Curies. From Table 1 we see that the contribution to the total fraction
! (0.188) of MFC at the site tvuWy due to Kr-85 after 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> is only
! 0.00009/0.188 = 4.8 x 10-4 At 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> exposure, this becomes only 9.6 x ,
10-4 or 0.0964, a totally negligible increase in the fraction of MFC released ,
j to the uncontrolled region beyond the site boundary. l I !
l Radiation Effects on strength of containment Camanents i
i j The radiation effects on significant regions of the fission product i j containment have been evaluated for a 40,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> irradiation in a fast neutron ,
flux of 2.9 x 1013 nv (>0.1 Mev) . The corresponding fluence of fast neutrons l ,
i is about 4.2 x 1021 nyt. The itms of concern are the stainless steel Type 304 i primary and secondary containment, the associated stainless steel welds, and f 1
- l the hard vacuum (cermic-tW) seal. It is shown that the maxima stress l on the containment for any normal or abnormal operation is about 675 psi in the 1 high radiation areas within the reactor core. At a point above the top grid ;
$ plate where the incore section of the thermionic devices enlarge to the 4-inch I dianeter section, the corresponding maximaan stress is about 870 psi. These j working values of stress are tiny campared to the ultimate tensile' strength and j yield strength for Type 304 stainless steel operating at 100-391oC. In addi- i
! tion, Type 304 stainless steel irradiated to 4.2 x 10 21 nyt (E>.1 Mev) still
{ retains most of its ductility as indicated by the curves for elongation. Mtile !
! it is clear that same degradation is expected for the ultimate strengths of the ,
' l j ocntairement, it is also clear that the maximan working stress is but a tiny j fraction of the capability of the containment and that the degradation will not j
] affect the safety of the fission product ocntainment. Additionally, it is j clear that the containment walls still retain most of the original ductility j giving assurance that the walls and welds will not break fece brittle fracture.
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Details of containment t
{ The direct conversion devices contain one or more thermionic energy converters within a contalment. The contaiwant consists of three independent i and sealed regions which are separated b" 4 hard vacuum (ceramic-to-metal) i seal. Below the hard vacuum seal, the. contalment consists of concentric
} primary and " lower" secondary regions wL8.ch are coepletely independent of each
! other. The primary contalment stops at the hard seal. The contaiment for ,
l the region above the hard vacuum seal consists of a cylindrical " upper" secondary region which is independent of both the primary and " lower" secondary regions. The upper secondary region is sealed at the top by a 30-inch long j epoxy plastic berrier plus another 15 inches of plastic and Pb0 glass barrier.-
l The lower containment portion containing the converters has outside dimeters of 1.475 or 2.341 inches.* A few inches above the top grid plate the contain- !
j ment enlarges to 4 or 2.875 inches in diameter. An overall sketch of a typical j device is shown in Figure 1. Details of the direct conversion devices (called j ??E on the sketch) indicated in Figure 1 are shown schematically in Figure 2. l l The incore fluence of fast neutrons 00.1 Mev) at the center of the thermionic l device is about 4.2 x 1021 nyt. A conservative estimate of the lowered fluence !
) at the lower end of the 4-inch section is about 6 x 1018 nyt. A conservative i estimate of the fluence at the position of the hard vacum seal ranges from 5 x 1017 to 1.8 x 1018 nyt depending upon whether a 1.477-inch or a'2.341-inch section is used.
a i
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) *these dimensions represent the outside dimeters for the two capsule desips ;
l which are in use or planned for the present progran. The two desips, however, l i have identical contalment configurations.
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._ _ . -_ ._ - __ . . . _ _ _ _ . _ . _ ~ _
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Safety of Type 304 stainless steel contaimant i
- '1he irradiation of the 304 stainless steel cm i.La and their welds
! occurs at relatively lw temperatures of 100 to 391oc. Data on Iw tesperature l irradiations (References 1 through 16) she that the values retained for strength and ductility of 304 stainless steel are more than sufficient to assure that the safety of the contaimants are not ocuptcmised upon extension l
of their exposures to 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> and 4.2 x 1021 (E >0.1 Mov) of operation.
'1he range of values found in the above references for the mechanical properties of irradiated 304 stainless steels and its welds are given in Table 2. '1he I effect of irradiation on 304 stainless steel at im tesperatures are-illus- ,
trated in Figures 3 through 7. Figure 3 shows the 0.2% yield strength vs l j neutron fluence; Figure 4 shes the values for ultimate tensile strength vs l neutron fluence; Figure 5 shows t reduction in area vs neutron fluence; Figure j 6 shows the change in t total elongation vs neutron fluence; and Figure 7 I
cagares the response of charpy impact ductility in ft-lbs vs neutron fluence for the 304 stainless base material and weldnents.
J l .
j Since the stresses under any conceivable condition of operation or i
4 handling are a very small fraction (<104) of the yield stress of irradiated 304 ;
stainless steel base metal or weld, no plastic deformation will occur and no
{ failure of the contalments can take place.
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Figure 1 Schanatic of TFE Model lH1
'IRIGA Test
- - WELD EMITTER BUS kEATTR LEADS (11) N STA llS.ll" caver g\ , , FtLL TUBE b \-HYSOL + DEVCou + FeBER BOARD GAS TutE 7, y g, #
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- STA 0" g N LoujER GRID PLATE Page 8
l Figure 2 Design Concept of Fueled Dnitter Deformation Capsule 1 ;
- 7.3 m (~24 FT)
INSTRUMENTATION
{0100E, HEATER, THERMOCOUPLE, 1 MANOMETER, ETC. LEADS)
=
0.9 m (~3 FT) MANOMETER
/ . 44 f P '9y / / I lxM W / RIMARY CONTAINMENT 4 \ \ \\ \ \ Y\\ \ \\\\\\ \\\\ \\ \ \ \ \ \1 h .\ \\\ \ \ \ \ \ \ \
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_ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ . .- _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ __ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ . . . . .~
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l Table 2. Iai Taperature Irradiation Effects on 304 Stainless Steel Base Metal and Weldments at 20,000 and 40,000 Hour Exposures (tw means No Data Found).
1 I 21 n/m 2D .lMev) 21 n/m 2 (D.lHev) l l 20,000 Hours (.75-2.1 x 10 1 40,000 Hours (1.5x-4.2x10 I i i I l l 1 1 I I l I som Tuperature 1 93 - 4500C I Rom Tenperature 1 93 - 4500C l References l 1 1 I I I i
i i i I i i i i l lI Base MetalI I Weld ii Base Hetal lI Weld I Base Metal I Weld l Base Metal I Weld I l ,o i l i i I
- 3 I I I I I I I I i
- '5 .2% Yield i I I I I I I I i
! ~ Strength (Est) I e i tw I 40-109 I tw I 56 1 71 1 38-109 I 56 l 1-8, 11-16 o i I I I I I I I I I I I I I I I I I Ultimate Tensile I I I I I I I I I Strength (Esi) I tw I IN I 83-110 1 }W I 97 8 98 I 67-110 1 73 I l-8, 11-16 l
l 1 i l I i i i i I i i i i i i i I l t El m gation i I I I l l ! I I l-6, 8, 1btal I 9-37 I IN I 3.3-24 1 }w I 9-21 1 31 1 3.3-21 l 16 l 10-16 I I I I I I l- 1 I I I I I I I I I I t moduction i .I I I I I I I I of Area i IN I IW l 30-35 I IN I }W I IN I 30-35 l IN I 2,4,6 I I I I I I I i l i i i l i l i i I Qarpy V-Notch l l l l 1 I i i l Ductility (ft-lbs) . l )W I IW l 86 .l 40 l IN I IW l 75 1 20 1 5,7,9,10 I i ! Based on IBased oni i I Based on Isased on!
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Another helpful point.can be made concerning the very low probability of brittle fracture. Se fact is that each thermionic device is densely packed ;
with delicate instrumentation and transducers and has closely controlled i internal spacings. Handling of the thermionic devices during moves associated ]
with neutron radiography (about every 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> of irradiation) nust of i necessity be very gentle. @is handling will result in very low mechanical stresses in the contaiment, far less than the yield strength of the base metal j as well as the welds. Further, any gentle jarring of the aaaambly that occurs during the gentle handling for NR will lead to internal inpacts that are a tiny fraction of the Charpy V-notch ductility (see Table 2) for both the base metal and welds. However, should a rupture of the primary occur, the fission products would still be contained in the secei-dary containnent. S e likelihood of a rupture of both the primary and secondary is even more remote. However,
! the subsequent release of fission products fra such a rare event would be l under 22 feet of water which would thoroughly scrub the gases of any iodine 2
l content. Without the contribution fra I-131 shown in Table 1, the dose to any workers within the reactor rocan would be mch reduced and the potential release to the site boundary would be auch smaller than evaluated therein. here is therefore no change in safety to the public due to embrittlement (loss of ductility) of the base metal nor is there any increase in the probability of a rupture.
Se strength of the Type 304 stainless steel containment as a function of irradiation is not in question because ==vi== stress in the containment base metal is but a tiny fraction of the tensile strength. % e stress in the wall is estimated fran the formula for a long thin wall tube:
on EE
- t where P is the pressure differential and I and i are the radius and thickness of the tube. For the condition yielding the maxinaan stress, the lower secondary cor*=i-mt tube is 0.027 inch thick and 1.475 inches in diameter.
l PB98 E l
In an evacuated pipe 22 feet under water, the max 2 mum stress is therefore, i l
l 24.7 x 0.7375 = 675 psi o=
4 0.027 For the 4-inch upper section of pipe in the lowered radiation field, the maximum stress is:
o = 870 psi.
We tenperature of the primary contaiment wall is estimated to be 3910C (7350F) . We temperature of the lower secondary (outer) contaiment is about that of the coolant water within the core, approximately 1040C (2200F) or lower. W e data in Figs. 3 to 7 cover the above temperature range.
For the incore sections, the maximum stress during normal or abnormal operation is only 675 psi and is far less than the yield strengths of more than 38,000 psi at 3910C listed for both base metal and welds. For the lower portion of the 4-inch rection where the maximtat stress in the base metal is 870 psi, the operating temperatures are on the order of 1000C and the fluence is reduced fra 4.2 x 1021 nyt to a value between 1.7 x 10 18 and 6.02 x 1018 nyt, depending on whether the incore section is 1.475 or 2.875 inches in diameter.
Wis low fluence assures that the mechanical strength and the ductility of the Type 304 base metal are far greater than the operational stress of 870 psi. In considering the effects of radiation on welds, it may be noted that the only welds of concern are located near the bott e and ap of the core region far from the peak neutron fluxes. A conservative estimate of 2 x 1021 nyt is appropriate for the 40,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> irradiation, compared to 4.2 x 1021 nyt for the central core region. As can be seen fra Figures 3 and 4, for example, the difference in tensile strength for 2 x 10 21 versus 4 x 1021 nyt lies within the scatter of data points.
Se hard vacuum (ceranic-to-metal) seal is located in a region where the operating temperature is less than, or on the order of,1000C (see Fig.1) and the reduced fluence of fast neutrons, at a maximum, lies in the range of 5.0 x 1017 to 1.8 x 1018 nyt. Wese low values are even lower if account is taken of Page 17
the large mass of intervening material between the incore location and the hard 7 vacutan seal. D e stainless steel portion of the ceramic-to-metal seal has no discernible radiation effects at these low fluences. The ceramic (Al23 0 ) has j been considered separately. Fra the published literature (Ref.17) the most j preinent radiation effect on Al 023 is volume ergansion (swelling). Se pertinent data presented in this reference is reproducec in Figure 8.- Se .
19 l lowest fluence with measured swelling is about 5 x 10 nvt at which the volume i increase is about 1.0%. Fra the clear trend we can conclude that no damage to the ceramic-to-metal seal will be caused by a fLience of about 10 18 nyt where the volume swelling is at most only a few tenths of a percent and the linear expansion would be roughly one third of this tiny value.
In sumary, fra the several considerations presented above on fission j product contalment, we conclude that only that part of the device in the
! incore location receives sufficient radiation to warrant attention. R1rther, the very low stress encountered in the actual system is so far below the
! residual strength of the materials that no decrease in safety of the public results from considerations of potential fission product release after irradiation for 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Similarly, the need to handle the device i carefully to avoid damage to the cmplicated internal instraentation provides
- added assurance that the slight increase in material enbrittlement will hav.e no substantial effect (from the point of view of brittle fracture) on the safety j of contaiment. As noted above, in the very unlikely event of a catastrophic i
rupture of both the primary and secondary contaiment, then the release of
. fission product gases would be at a release point under about 22 feet of water i and would create far less hazard than that which was analyzed earlier in this at*% for the rupture in air and which was found acceptable.
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i i l i
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REFERENCES i
l
- 1. Holmes, J.J. , et al, " Post Irradiation 'Innsile Behavior of 300 Series Stainless Steels," Conf-68060407, BWIr-SA-1455, June 1968.
- 2. Claudson, T.T. , et al, "The Effects of Fast Flux Irradiation on the Mechanical Properties and Dimensional Stability of Stainless Steel,"
Nuclear Appl & Tech, Vol. 9, Jtily 1970.
- 3. Weir, J.R. , et al, " Irradiation Behavior of Cladding and structural Materials," Proceedings of ANS Topical Meeting, April 2-4, 1968.
- 4. Conner, J. G. and Pormbka, S.W., "A Cmpendim of Properties and Characteristics for Selected LMFBR Cladding Materials," BMI-1900, May 15, 1968.
- 5. Hawthorne, J.R. and Watson, H.E., " Notch Toughness of Austenitic Stainless Steels with Nuclear Irradiation," Welding Research Supplement, pages 255-260, June 1973.
- 6. Ward, A.L. and Holmes, J.J. , " Ductility Loss in Fast Reactor Irradiated Stainless Steel," Nuclear Appl & Tech, Vol 9, November 1970.
- 7. Loss, F.J. and Gray, R.A., Jr. , "J-Integral Characterization of Irradiated Stainless Steels," NRL Beport 7565, April 25, 1973.
- 8. Holmes, J.J., and Straalsund, J.L., " Effects of Fast Neutron Exposure on the Mechanical Properties of Stainless Steels," Radiation Effects on Breeder Reactor Structural Materials, June 10-23, 1977, Scottsdale, Arizona.
- 9. Hawthorne, J.R. , and Watson, H.E. , " Exploration of the Influence of Welding Variables on Notch Ductility of Irradiated Austenitic Stainless Steel Welds," Radiation Effects on Breeder Reactor Structural Materials, June 10-23, 1977, Scottsdale, Arizona.
- 10. Van der Schaal, B., et al, " Irradiation Enbrittlement of Type 304 Stainless Steel Welds and Plate at 823K (5500C)," Radiation Effects on Breeder Reactor Structural Materials, June 10-23, 1977, Scottsdale, Arizona.
- 11. BE L Staff, "The Effect of Test Temperature on the Ductility of AISI 304 stainless Steel in a Flowing Sodim Enviroment," BMIr-710.
- 12. Blom, E.E., and Steigler, J.O., " Post Irradiation whanical Properties of Types 304 and 304 5.15% Titania Stainless Steel," Nuclear Technology, Vol.17, Jan 1973.
- 13. Blom, E.E., and Steigler, J.O., "Effect of Neutron Irradiation on the Ductility of Austenitic Stainless Steel," Nuclear Technology, October 16, 1972. l 1
i Page 20
REFERENCES (CONTINDE)
- 14. Carlander, R., et al, " Fast Neutron Effects on Type 304 Stainless Steel,"
Nuclear Appl. & Tech., Vol. 7, July 1969.
- 15. Holmes, J.J., et al, "Effect of Fast Reactor Irradiation on the Tensile Properties of 304 Stainless Steel," Jour. of Nuclear Materials 32, pp 330-339 (1969).
- 16. Dufresne, J., et al, " Fracture Toughness of Irradiated AISI 304 and 316L Stainless Steels," Effects of Radiation on Structural Materials, ASIM STP683, pp 511-528, 1979.
- 17. Keilholtz, G.W., et al, " Fast-Neutron Damage to Polycrystalline Alumina at Temperature frcxn 60 to 12300C," Nuclear Technology, Vol.17, pp 234-246, 1973.
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